ML20078M942
| ML20078M942 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/01/1994 |
| From: | Mckee P Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20078M946 | List: |
| References | |
| NUDOCS 9412050301 | |
| Download: ML20078M942 (5) | |
Text
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UNITED STATES y-4 7
E NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. N1
%,.....,o GPU NVCLEAR CORPORATION AHQ JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. s 75 License No. DPR-16 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by GPU Nuclear Corporation, et al.
(the licensee), dated June 24, 1994, as supplemented September 30, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is re.sonable assurance (i) that the activities authorized by this rnendment can be conducted without endangering the health and sa'ety of the public, and (ii) that such activities will be cond'.cted in compliance with the Comission's regulations; l
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and i
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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9412050301 941201 DR ADOCK 0500 9
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2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-16 is hereby amended to read as follows:
(2)
Technical enecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.175, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
t 3.
This license amendment is effective as of the date of issuance, to be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/
Cf Phillip F. McKee, Director Project Directorate I-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: December 1, 1994 i
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ATTACHMENT TO LICENSE AMENDMENT NO.175 FACILITY OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert 2.3-3 2.3-3 2.3-7 2.3-7
FUNCTION LIMITING SAFETY SYSTEM SETTINGS K.
Reactor Low-Low Water Level, 17'2" above the top of the Core Spray Initiation active fuel L.
Reactor Low-Low Water Level, 17'2" above the top of the Isolation Condenser Initiation active Fuel with time delay 53 seconds M.
Turbine Trip, Scram 10 percent turbine stop valve (s) closure from full open N.
Generator Load Rejection, Initiate upon loss of oil Scram pressure from turbine acceleration relay 0.
DELETED P.
Loss of Power 1) 4.16 KV Emergency Bus 0 volts with 3 seconds t Undervoltage (Loss of 0.5 seconds time delay Voltage) 2)
4.16 KV Emergency Bus 3840 (+20V, -40V) volts Undervoltage (Degraded 10 10% (1.0) second time Voltage) delay Bases:
Safety limits have been established in Specifications 2.1 and 2.2 to protect the integrity of the fuel cladding and reactor coolant system barriers, respectively.
Automatic protective devices have been provided in the plant design for corrective actions to prevent the safety limits from being exceeded in normal operation or operational transients caused by reasonably expected single operator error or equipment malfunction. This Specification establishes the trip settings for these automatic protection devices.
The Average Power Range Monitor, APRM"', trip setting has been established to assure never reaching the fuel cladding integrity safety limit. The APRM system responds 1
i to changes in neutron flux.
However, near the rated thermal power, the APRM is calibrated using a plant heat balance, so that the neutron flux that is sensed is read out as percent of the rated thermal power. For slow maneuvers, such as those where core thermal power, surface heat flux, and the power transferred to the water follow the neutron flux, the APRM will read reactor thermal power.
For fast transients, the neutron flux will lead the power transferred from the' cladding to the water due to the effect of the fuel time constant. Therefore, when the neuron-flux increases to the scram setting, the percent increase in heat flux and power transferred to the water will be less than the percent increase in neutron flux.
The APRM trip setting will be varied automatically with recirculation flow, with the trip setting at the rated flow of 61.0 x 10' lb/hr of greater being 115.7% of rated neutron flux.
Based on a complete evaluation of the reactor dynamic performance during normal operation as well as expected maneuvers and the various mechanical failures, it was concluded that sufficient protection OYSTER CREEK 2.3-3 Amendment No. /A, /5, 80, 174, 175
i valves to a load rejection and failure of the turbine bypass system. This scram
-is-initiated by the loss of turbine acceleration relay oil pressure. The timing
'for this scram is almost identical to the turbine trip.
l The undervoltage protection system is a 2 out of 3. coincident logic relay system designated to shift emergency buses C and D to on site power should normal power be lost or degraded to an unacceptable level.
The trip points and time delay i
settings have been calected to assure an adequate power source to emergency safeguards systemt in the event of a total loss of normal power or degraded conditions which voaid adversely affect the functioning of engineered safety features connected tc_the plant emergency power distribution system.
References i
1 (1)
FDSAR, Volume I, Section VII-4.2.4.2 (2)
FDSAR, Amendment 28, Item III. A-12 (3)
-FDSAR, Amendment 32, Question 13 (4)
Letters, Peter A. Morris, Director, Division of Reaction Licensing, USAEC to John E. Logan, Vice President, Jersey Central Power and Light
- Company, Dated November 22, 1967 and January 9, 1968 (5)
FDSAR, Amendment 65, Section B.XI.
(6)
.FDSAR, Amendment 65, Section B.IX i
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i OYSTER CREEK 2.3-7 Amendment No.: /6, gd,175
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